ML20070L685

Application for Amend to License NPF-85,consisting of Tech Spec Change Request 90-16-2,changing TS Sections 2.1.2, 3.2.1 & 3.2.3 Re MCPR Safety Limit/Revised Calculational Methods Per Changes in Reload Fuel Type

ML20070L685
Person / Time
Site:	Limerick
Issue date:	03/12/1991
From:	Beck G PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20070L688	List: ML20070L685 ML20070L689
References
NUDOCS 9103200160
Download: ML20070L685 (9)

Text

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10 CFR-50.90 PHILADELPH.A ELECTRIC COMPANY-NUCLEA' GROUP HEADQUARTERS .

iro5 65 CHESTERBROOK BLVD .

WAYNE PA 19087 5691 (sis) sao.sooo D)cket No. 50-353 License No. NPF-85 U. S. Nuclear Regulatory Cminission ATIN: Document Control Desk Washiston, D: 20555 SUBITT: Limerick Generating Station,' Unit 2 Technical Specifications Change Request Gentlemen:

' Philadelphia Elect.ric Cm' is sutznitting Tec.hnical Specifications Change ' '

. Request No. 90-16-2, in accordance with 10 CFR 50.90, requestig an _ ameJrtment- to +

the Technical Specifications (TS) (Apperrlix A) of Operating License'No. NPF-85.

Infomation supportirg this Change Request is contained in-Attachment 1 to this letter, and the proposed TS replacement pages are contained in Attachment 2.'

{ This subnittal requests changes to TS Sections 2.1.2, 3.2.1, 3.2.3, and the pertinent TS Bases to reflect chaNes .to the Minimum Critical Power

, Ratio (MCDR) Safety Limit as a result of changes in reload fuel cype and to S

! reflect the revised v:xnputer methods used to calculate thermal limits.

l If you have any-questions regarding this matter, please contact us.

l j Very truly yours, M ~

i. G. J. Beck

Manageri Licensing j Nuclear Engineering-and Services .

i GHS/eas:3033

, Attachments cc: T. T. Martin, Administrator, Region I, USNRC-4 T. J l<enny, USNRC Senior Resident Inspector, IGS j- T. M. Gerusky, Director, PA Bureau of Radiological Protection 4

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CCtNotMEALTH OF PDNSYIVANIA :

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COUtTIY OF PHIIADELPHIA  :

G. M. Initch, being first duly sworn, deposes arxl says:

'Ihat he is Vice President of Itiladelphia Electric Coopany; the Applicant herein; that he has read the foregoing Application for Amerdment of Facility Operating License No. NPF-85 ('Ibchnical Specifications Change Request No. 90-16-2) to reflect changes to the Minimum Critical Power Ratio (MCPR) Safety Limit as a result of changes in reload fuel type and to f

reflect the revised computer methods used to calculate thermal limits, the contents thereof; and that the statements and matters set forth therein are true and correct to the best of his knowledge, infonnation and belief.

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8 ( .V Vice President

Subscribed and sworn to before me this t6" day of f 'insm 1991, i

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' Notary Public NOTAPlALSS\L

, OLT TOW FRANKLIN. N":y M";

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ATIAONDfr 1 LIMDUCK GDIDETIIC STATION Unit 2 Docket No. 50-353 Liccatso No. NPF-85 TIX1NICAL SPECIFICATIONS QWGE REQUEST "Mininnta Critical Power Ratio Safety Limit / Revised Calculational Methods Changes" Supporting Informtion for Changes - 6 pages

4 Attachntent 1 Page 1 Philadelphia Electric Canpany (PECo), Licensee under Facility Operating License NPF-85 for Limerick Generatirg Station (LCS), Unit 2, rcquests that the Technical Specifications (TS) containod in Appendix A of the Operating License be amended as proposed herein to reflect changes to the Minimum Critical Power Ratio (MCPR) Safety Limit as a result of chmges in reload fuel type aM to  ;

reflect the revised corputer methods used to calculate thermal limits as previously approved by the NRC. Se proposed 7S dianges are indicated by a vertical bar in the mrgin of 'IS pages xviii, 2-1, 3/4 2-1, and 3/4 2-8, and Bases pages B 2-1, B 3/4 2-1, B 3/4 2-2, B 3/4 2-3, and B 3/4 2-5 for LGS, Unit 2, and are contained in Attachneat 2.

We request the changes p mpo d herein be effective upon issuance of the Anendment.

This Change Request provides a discussion and description of the proposed TS changes, a safety assessment of the proposed TS changes, informtion 4 supporting a finding of No Significhnt Hazards Consideration, and information suppor*ing an Envimnmental Assecsment.

Dimission _and Descriotion of the Proposed Cb3D9eJ Implementation of this Qiange Request involves the proposed TS changes

described below, i

! We first proposed TS change would revise the current MCPR Safety Limit of 1.06 to a new limit of 1.07 for two recirutlation loop operation (1.07 to 1.08 for single recirculation loop operation) . Bis increased limit would provide additional conservatism to account for possible uncertainties in power i distribution in reload reactor cores, he proposed MCPR Safety Limit is in accordance with Revision 9 of " General Electric Standard Application for Reactor i Fuel," (CESTAR-II), NEDE-240ll-P-A-9, September 1988 (Reference 1) and was j apptrved by the NRC in a letter from Ashok C. Thadani (NRC) to J. S. Charnley 4

(GE), " Acceptance for Referencing of Amendnent 18 to General Electric Licensing Topical Report NEDE-24011-P-A-9, ' General Electric Standard Application for

, Reactor Fuel'," dated May 12, 1988 (Reference 4).

i %e second proposed TS change would substitute new computer vl: hods I

(i.e. , GmINI/ODYN) for the current conputer methods (1.c. , GDIESIS/ODYN) as the

'. basis for calculating therml limits. TS changes to reflect the GmINI/ODYN j

computer methods and multiple lattice fuel are proposed to the Average Planar Linear Heat Generation Rate (APUlGR) Bases Section 3/4.2.1, the APUlGR Limiting

Condition for Operation (If0) 3.2.1, and the MCPR ILO 3.2.3. These changes are proposed in accordance witn Revision 9 to GESTAR-II and were approved by the NRC in References 5 and 7. These proposed changes are identical to the current LGS Unit 1 TS as previously approved for Unit 1 by the NRC (see Reference 6) .

Below is a discussion of the propcFM TS changes.

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1 Attachment 1 Page 2 l

MCPR SAFf'N LIMIT The fuel cladding integrity safety limit is set such that no fuel damage is calculatcd to occur if the limit is not violated. Since the parametere which result in fuel damge are not directly observable during reactor operation, the therml hydraulic corditions resulting in a departure from nucleate boilirg have been used to mark the beginning of the region where fuel damaoe could occur.

Although it is recognized that a departure fram nucleate boiling would not necessarily result in damage to Boilire Water Reactor (IMR) fuel rods, the critical power et which boiling transition is calculated to occur has.been adopted as a convenient limit. However, the uncertainties in monitoring the core operating state and in the procedure used to calculate the critical power result in an uncertainty in the value of critical power. Therefore, the fuel cladding integrity safety limit (i.e. , the MCPR Safety Limit) is defined as the critical power ratio on the limiting fuel assembly for which more than 99.9% of the fuel rods in the core are e.xpected to avoid boiling transition considering the power distribution within the core and all uncertainties.

7he MCPR Mfety Limit is detel'nined using the NRC approved General Electric Thermal Analysis !bsis (GETAB) described in Referen as 1 ard 2 for two recirculation loop operation. 7he MCPR Safety Limit is increased by 0.01 for single recirculation loop operation as discussed in Reference 3.

A MCPR Safety Limit of 1.07 for two recirculation loop operation (1.08 for single recirculation loop operation) has been previously approved by the NRC for application to C-lattice plants operating with a reload reactor core of GE8x8NB (i.e., GE9B) fuel (see Reference 1, Table 4-2). ICS Unit 2 is a C-lattice plant and the reload fuel for the second cycle of opemtion will be GE9B fuel, with the exception of twelve Qualification Fuel Burdles (QFBs). Howver, these QFBs will be loaded in non-limiting locations as descrilxxl in our letter to the NRC dated March 11, 1991, such that the QFBs will have an insignificant impact on the core wide MCPR Safety Limit.

( GDtINI/ODYN COMEUTER ME'IHODS Amendment 11 to GESTAR-II revises the docunent to include an updated version of the ODYN computer code amory the calculational techniques used for plant transient analyses and updates the manner in which calculational uncertainties are treated in obtainirg core operating limits. ' Ibis new approach l is the GEMINI /ODYN computer methodology and is similar to the previously l approved GENESIS /ODYN computer methodology. The changes to the ODYN calculational mortel include:

1. improved neutronics methods,

2. inclusion of GESTR-M Riel Performance Model,

3. improved Bulkwater Model

4. improved Upper Plenum Model, and

5. improved Steam Separator Mass Storage Model.

m Attachment 1 Page 3 Data provided to the NRC on the results of comparisons of the old and new ODW conputer code calculations for the Peach Dottcn M antic Power Station turbine trip tests showed that the new ODYN results (1.9., GEMINI /ODW) provide generally better agreement with the test data than did the old ODYM (i.e. , GDESIS/ODW) calculations. Due to the increased accuracy of the new ODW calculations and a revised mnner in which computer code uncertainties are <

handled in obtainirq the Option A and Option B MCW operating limits, the Statistical Adjustment Factors (SAFs) used in obtaining the MCPR operating limit 95/95 value (i.e., that vr.lue which assures a 95% probability with 95%

confidence that the critical pvn will not fall below the MCPR Safety Limit) for pressurization events have c..angad. These SAFs result in improvements to the 95/95 MCPR operating limits while mintaining the same margin of safety.

A separate change that also affects the MCPR operating limit is an expansion of the dat:"ase used for detemining the scram speed distribution (i.e., used to derive the analytical control rod insertion time). This results in new values of the mean and stardard deviation of the scram speed database.

These values are used to detemine the value of TAU as shcun in prow TS Section 3.2.3. The calculation of TAU is used to demonstrate that the ICS Unit {

t 2 scram speed distribution is consistent with that used in the calculation of the 95/95 MCPR operating limit. If scram speed is not demonstrated by test to comply with that used in the 95/95 MCPR operating limit calculation, then a more conservative MCW operating limit that does not include full credit for the statistically evaluated scram times must be used. The determination of the mre 1 conservative MCPR operating limit is done in the sam way that it is currently l done with the exception of the new values for the mean and standard deviation of the scram speed database.

01anges in the wording in the TS Bases and the TS LCO for the APUlGR and in the TS ICO for MCPR is also being proposed. These proposed 15 charges are administrative in nature and reflect the GD4INI/ODYN computer methods and multiple lattice fuel.

Safety Assessment The MCPR Safety Limit is set such that no fuel damage is calculated to occur if the limit is not violated. IGS Unit 2 is a C-lattice plant and the reload f'lel for the second cycle of operation will be GE9B fuel. The pro W MCPR Safety Limit was determined using NRC approved methodology for C-lattice plants operatirg with GE9B reload fuel. *he use of this methodology ensures that the proposed MCPR Safety Limit m*' wins the same level of conservatism regarding calculational uncertaintia, and therefore, the same mrgin of safety as that for the current fuel. Operation of the plant based on the proposed MCPR Safety Limit will ensure that fuel cladding integrity is mintained. The proposed changes to the APUlGR TS Bases, APUlGR TS ILO, and MCPR TS ILO reflect the use of this revised NRC approved methodology for calculating themal limits, which ensures that the same level of conservatism is maintained. The proposed 1S changes are to analytical values and methods, and do not physically affect the fuel. Mditionally, the proposed TS changes do not alter the design or function of plant equipment, nor do they introduce any new operatirq scenarios, configurations, or fellure modes.

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Attachment 1 Page 4 Informtion Succortina a Findina of No Sicinificant Hazartis ConsideratigD We have concluded that the proposed changes to the IDS TS, which revise the MCPR Safety Limit to reflect the reload fuel type and the APulGR IID, MCm 140, and APUiGR Bases to reflect the revised methodology for calculatirs themal limits, do not constitute a significant Hazards Consideration, ln support of this determination, an evaluatton of each of the thIre standartls set forth in 10l CFR 50.92 is provided below.

1. %e protosW chames do not involve a sictnificant increase in the probability or conseauences of an accident nrevious1v evaluated.

%e proposed changes to the MCPR Safety Limit ard the themal limit calculational methods are charges to analytical values and methods, and, in themselves, cannot initiate an accident. %e MCPR Safety Limit is set such that no fuel damage is calculated to occur if the limit is not violated and is determined' bared on the revised NRC apptuved methodology. ' % e use of this muthodology ensures that--the same level of conservatism is maintained with respect to calculational--

uncertainties. Operation of the plant based on the proposed MCPR Safety Limit will ensure that fuel cladding integrity is maintained.

% erefore, the proposed TS changes would not cause an increase in the-probability or consequences of any accident previously evaluated.

2. The ottoosed chances do not .c mate the oossibility of a new or l different kind of accident iram any accident previously evaluated.

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%e proposed TS changes are to analytical values and methods and do not physically affect the fuel,-and therefore, in themselves, cannot initiate any accident or cause any type of fuel malfunction. The proposed TS changes do not alter the design or function of any plant equipment, nor do they introduce any new operating scenarios, configurations, or failure modes that would create the-possibility of

- a new or different kind of accident from any accident previously evaluated.

3. We procosed chames rio not involve a sianificant reduction-in a mrcrin of safety.

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%e martfi n of safety is based upon the methods used to dotermine the MCPR Safety Limit and the other thermal limits. The proposed TS changes are to the value of' ne MCPR Safety Limit as determined by these methods and to the calculational methods themselves as reflected in the proposed APLHGR TS ILO and Bases,-and the MCPR TS IfD. Thesc-changes have been reviewed ard approved by .the-NRC and will maintain the same martf i n of safety.

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Attachmnt 1 Tbge 5 I

l Informtion Supnortim an Environmntal Assessment l

An environmntal assessment is not required for the changes pIrposcd by this 01arge Prquest because the roauested changes conform to the criteria for

" actions eligible for categorical exclusion," as specified in 10CFR51.22(c) (9) .

The requested changes will have m impact on the envircrant. The requested charges do not involve a significact hazanis consideration as discussed in the preceding section. The requested Avges do not involve a significant change in the types or significant increase in te amounts of any effluents that my be released offsite. In addition, the pr g d charges do not involve a significant increase in individual or cumulative occupational radiation exposure.

l Conclusion

( The Plant Operations Review Committee and the Nuclear Review Boatti have i reviewed these proposed changes to the TS and have concluded that they do not involve an untyviewed safety question, or a significant hazarris consideration, and will nec erdanger the health and safety of the public, i

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1. t Attachment 1 Ibge 6 1

References i

1. " General Electric Staniard Application for Reactor Fuel," NEDE-24011-P-A-9, September 1988,

2. " General Electric IMR 'Ihermal Analysis Basis (GEFAB): -!Mta, Correlation and Design Application," NED}-10958-A, January 1977.

3. " Single loop Operation Analysis for Limerick Generatim Station Units 1,"  !

NEDC-31629P, Aaotember 1988- -

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l 4. Ashok C. Thadani (NRJ)' to J. S. Charnley (GE), " Acceptance for Refelweing of Anentnent 18 to General Electric Licensing Topical Report NEDE-240ll-P-A-9, ' General Electric Standard Application for Reactor.

Fuel'," dated May 12, 1988.

5. G. C. Lainas (NRC) to J. S. Charnley -(GE) " Acceptance for Referencing of -

Lice wing 7bpical Report NEDE-24011-P-A, _ 'GE Generic Licensing Reload

- Report,' Supplement to Amendment 11," dated March _22, _1906.-

6. Safety Evaluation _by the Office of Nuclear Reactor Regulation, NRC, supportig Amendment No. 7 to Facility Operating License No. NPF-39, .

Fhiladelphia Electric Company, Limerick Generating Station, Unit 1, Docket No. 50-352, dated August 14, 1987.

7. Ashok C. 'Ihadani (NRC) to J. S. Charnley -(GE),- " Acceptance for Referencing _

l -of Amendment 19 to General Electric' Licensing 7bpical- Report-NEDE-24011-P-A-9, ' General Electric Standard Application for Reactor Fuel',

dated April 7, 1987," November 17, 1987.

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